1. Field of the Invention
The present invention relates to the field of deposition of metals, especially electrodeposition of actinides, and especially the room temperature electrodeposition of lanthanides and actinides from ionic liquids.
2. Background of the Art
Reclaiming unspent nuclear materials, while separating and sequestering fission products is extremely important in the management of the growing stockpile of nuclear waste. More importantly, reclamation of the actinide metals is important for future safety due to the possible proliferation of weapons. (Morss, L. R.; Edelstein, N. M.; Fuger, J.; Katz, J. J.; Kirby, H. W.; Wolf, S. F.; Haire, R. G.; Burns, C. J.; Eisen, M. S. The Chemistry of the Actinide and Transactinide Elements; third.; Springer Netherlands, 2006) Finally, reclamation of unused uranium from nuclear fuel is of general importance for reuse in energy processes and for the production of target material to generate useful radio pharmaceutical species for biological applications. (Hofman, G. L.; Wiencek, T. C.; Wood, E. L.; Snelgrove, J. L.; Suripto, A.; Nasution, H.; Amin, D. L.; Gogo, A. In 19th International Meeting on Reduced Enrichment for Research and Test Reactors; 1996.)
Typical electrochemical processes to recover uranium from spent nuclear fuel result in the accumulation of minor actinides (americium (Am) and curium (Cu)) and transuranic elements (plutonium (Pu) and neptunium (Np)). These accumulated elements usually occur as metal chlorides in the molten electrolytic salt. They must periodically be removed from the electrolyte for the fuel reprocessing to continue.
The simplest method to recover the target elements is via chemical or electrochemical reduction. Electrochemical reduction has two advantages over chemical reduction. The first advantage is that the site of reduction is localized to the cathode surface forming a cathode deposit affording easy removal from the process equipment. The second advantage is that the use of electrons as the reducing agent does not add to the waste volume. Deposition of the transuranic elements and minor actinides on a solid cathode is well-known. Accompanying anode reactions include the oxidation of chloride ions to chlorine gas, oxidation of a sacrificial alloy, and oxidation of metallic uranium or reduced light water reactor (LWR) feed material.
U.S. Pat. No. 7,267,754 (Willit) discloses an improved process and device for the recovery of the minor actinides and the transuranic elements (TRU's) from a molten salt electrolyte. The process involves placing the device, an electrically non-conducting barrier between an anode salt and a cathode salt. The porous barrier allows uranium to diffuse between the anode and cathode, yet slows the diffusion of uranium ions so as to cause depletion of uranium ions in the catholyte. This allows for the eventual preferential deposition of transuranics present in spent nuclear fuel such as Np, Pu, Am, Cm.
U.S. Pat. No. 6,233,298 (Bowman) describes a subcritical reactor-like apparatus for treating nuclear wastes, the apparatus comprising a vessel having a shell and an internal volume, the internal volume housing graphite. The apparatus has means for introducing a fluid medium comprising molten salts and plutonium and minor actinide waste and/or fission products. The apparatus also has means for introducing neutrons into the internal volume wherein absorption of the neutrons after thermalization forms a processed fluid medium through fission chain events averaging approximately 10 fission events to approximately 100 fission events. The apparatus has additional means for removing the processed fluid medium from the internal volume. The processed fluid medium typically has no usefulness for production of nuclear weapons.
Uranium Separation Process, U.S. Pat. No. 3,030,176, April 1962. This work outlines the dissolution of Uranium and the separation of species from fission products. The work outlines the use of molten salts in the separation. The advantage of our method is that RTIL solutions are ionic providing the same properties without the need for elevated temperatures (500-750 C) that the molten salts require which reduces the production of unwanted gases in the recovery process.
Electroseparation of actinide and rare earth metals. U.S. Pat. No. 5,582,706, Dec. 10, 1996. The work outlines a pyrochemical process used to recover 99% of the transmutable fission materials. The process uses the electrochemical separation of the waste component. Our method does not require multiple paths or pyrochemical methods to achieve dissolution or separation of the fission products.
Actinide Dissolution. U.S. Pat. No. 5,205,999, Apr. 27, 1993. The work documents the dissolution of the actinide and lanthanide species in aqueous solution between pH 5.5 to 10 utilizing complexing agents. Our methods are conducted under similar conditions in room temperature ionic liquid. The same solution is used for the electrochemical separation and deposition of actinide species. Our method does not require complexing agents, it is performed in non-aqueous solution, and the direct dissolution is achieved in the same solvent system used for electrodepositon.
Magnesium Reduction of Uranium Fluoride in Molten Salts. U.S. Pat. No. 4,552,588, Nov. 12, 1985. The work documents the use of Mg molten salts in the reduction of UF4 to U metal. The temperatures required for the process are on the order of 1000 degrees. There are inherent dangers associated with molten salts at high temperatures that are eliminated when RTIL solutions are used.
To date the PUREX process is the most widely utilized methods for the reclamation of actinides (Uranium and Plutonium) from partially spent nuclear materials. PUREX is an acronym standing for Plutonium—URanium EXtraction—the standard aqueous nuclear reprocessing method for the recovery of uranium and plutonium from used nuclear fuel. It is based on liquid-liquid extraction ion-exchange. The PUREX process was invented by Herbert H. Anderson and Lamed B. Aspreyas part of the Manhattan Project. Their U.S. Pat. No. 2,924,506, “Solvent Extraction Process for Plutonium” filed in 1947, mentions tributyl phosphate as the major reactant which accomplishes the bulk of the chemical extraction.
The method utilizes a complexing agent, tri-n-butylphosphate (TBP) and organic solvent such as kerosene or n-dodecane in the extraction and reclamation process. Modifications to the process have been primarily focused on developing new complexing agents or using different solvents for extraction. More recently the RTIL solutions have been examined as an alternative to more volatile organic diluents using tricaprylmethylammonium thiosalicylate as the complexing agent in the extraction of U into RTIL solution. Srncik, M.; Kogelnig, D.; Stojanovic, A.; Koerner, W.; Krachler, R.; Wanner, G. Uranium extraction from aqueous solutions by ionic liquids, Applied Radiation and Isotopes (2009), 67(12), 2146-2149. The added benefit of RTIL solutions is that it can be used in the direct electrochemical deposition of lanthanide or actinides species due the large potential window afforded by the non-aqueous system.
At present the accepted electrochemical method utilized to obtain uranium metal is based on molten salt eutectic system. (Iizuka, M.; Koyama, T.; Kondo, N.; Fujita, R.; Tanaka, H. Journal of Nuclear Materials 1997, 247, 183-190. Kim, K. R.; Bae, J. D.; Park, B. G.; Ahn, D. H.; Paek, S.; Kwon, S. W.; Shim, J. B.; Kim, S. H.; Lee, H. S.; Kim, E. H.; Hwang, I. S. J Radioanal Nucl Chem 2009, 280, 401-404. Koyama, T.; Iizuka, M.; Shoji, Y.; Fujita, R.; Tanaka, H.; Kobayashi, T.; Tokiwai, M. Journal of Nuclear Science and Technology 1997, 34, 384-393.
Internationally there are two well developed molten salts processes for the reprocessing/waste conditioning of irradiated nuclear fuel. A process developed by the Dimitrovgrad SSC-RIAR process uses high temperature (1000K) eutectic molten salt mixtures as solvents for the fuel and also as electrolyte systems. In this Russian system the solvent is typically an eutectic mixture of NaCl/KCl or CsCl/KCl. The process uses chemical oxidants (chlorine and oxygen gases) to react with powdered UO2 fuel, or mixtures of UO2 and PuO2, to form higher oxidation state compounds such as UO2Cl2 which are soluble in the molten salt. At the cathode the uranium and, if applicable, plutonium compounds are reduced to UO2 or UO2—PuO2, which form crystalline deposits. However, after a period of use the molten salt becomes loaded with fission products which not only begin to affect the quality of the product, but also result in too much heat generation within the salt. These fission products are commonly, but not exclusively, highly active lanthanide or actinide elements which may need to be isolated in a suitable form for immobilisation as a waste.
In the process developed by Argonne National Laboratory (ANL) in the USA, molten LiCl/KCl eutectic mixtures containing some UCI.sub.3 are generally used, rather than systems containing sodium or caesium salts, and a high temperature (around 773K) is again employed. However, single salts, such as LiCl, are suitable if higher temperatures are required, for example in the electrochemical reduction of fuel oxides. The process treats the spent nuclear fuel by flowing a current to oxidize a uranium anode and form uranium ions in the molten salt electrolyte. At the cathode the uranium is reduced and deposited as uranium metal. The ANL process is, unfortunately, a batch process, since the uranium is collected in a receptacle at the bottom of the apparatus, requiring that the process is interrupted in order that the receptacle may be withdrawn and the product recovered. In addition, the operation of the process is mechanically intense, involving the use of rotating anodes which are designed to scrape the product off the cathodes; difficulties are encountered on occasions due to the seizure of this mechanism.
While these methods have been utilized to produce U metal, it is not without flaws. From an engineering standpoint, the high temperatures needed for a molten salt system create safety and cost issues for the vessel material fabrication. (Avallone, E.; Baumeister, T.; Sadegh, A. Marks' Standard handbook for mechanical Engineers; 11th ed.; Mc-Graw Hill Professional, 2006. Creep & Fracture in High Temperature Components: Design & Life Assessment Issues; Shibli, I.; Holdsworth, S.; Merckling, G., Eds.; DESTech Publications, Inc., 2005). In addition, gas evolution is problematic due to environmental concerns and the safety of the workers. The second method is based on the synthesis of UF4 using HF gas. (Pushparaja; Poplit, K.; Kher, R.; Iyer, M. Radiation protection dosimetry 1992, 42, 301-305.) The process is expensive and dangerous process due to the health hazards and corrosive nature of hydrofluoric acid. In addition, reduction of the UF4 to metal using plasma and hydrogen is complicated by disproportionation and production of UF3 limiting the overall metal conversion.
All references cited herein are incorporated in their entirety by reference.